It will be appreciated that the fuel assembly of a light water nuclear reactor of the pressurised water type (PWR) or boiling water type (BWR) is composed in particular of a bundle of “rods”, that is to say, claddings containing pellets based on uranium oxide constituting the fuel, and various structural components bringing about positioning and mechanical retention of the rods and the handling thereof (guide tubes, retention grids, springs, housing, water duct, etc.). The claddings and at least some of the structural elements are constructed of zirconium alloy, that element having the advantage of allowing neutrons to pass, and also of being able to have the mechanical properties and corrosion resistance necessary in the media involved when the alloy elements and their contents are advantageously selected.
Among the zirconium alloys conventionally used, the following non-exhaustive examples may be set out:
alloys which are referred to as “Zircaloy” and which contain in particular tin, iron, chromium, oxygen and, in some cases, nickel;
Zr—Nb alloys such as M5® from the company AREVA NP, which contain approximately from 1 to 2.5% by weight of niobium, or also other elements, in particular such as oxygen, iron and tin.
Those compositions are given only by way of example, as an indication of one of the preferred contexts of the invention, it being applicable to zirconium alloys of all precise compositions, and also hafnium alloys, as will be seen below.
Similarly, it will be appreciated that, in order to adjust the reactivity of the core of the nuclear reactor during operation of the reactor, the elements absorbing the neutrons are moved inside (PWR reactors) or between (BWR reactors) some assemblies of the core in a vertical position, either in the insertion direction or in the removal direction, so as to introduce a greater or lesser length of the absorbent portions into the assemblies of the core. The control rods for pressurised water reactors are constituted by a support which is generally designated “spider” and a bundle of absorbent rods which can slide inside the guide tubes of the assembly. They are referred to as control clusters.
An absorbent rod can be constituted by a tube of hafnium optionally containing another absorbent material, or a bar full of hafnium.
The control rods of a boiling water reactor are generally in the form of plates which are assembled in cross-like form and are capable of sliding between the housings of the fuel assemblies. The plates may be of an absorbent material (hafnium), or be of steel and comprise inserts of absorbent materials (for example, B4C or hafnium). They are generally referred to as control crosses.
Hafnium is the metal absorbing neutrons originally used for control clusters of civil and naval reactors. Although subsequently reserved for military use, today it is increasingly used as a neutron absorbent in light water reactors, its availability having increased greatly with the development of zirconium alloys. Hafnium is present at a ratio of from 2 to 4% in zirconium ore, and must be separated therefrom owing to its great neutron absorption. Therefore, it constitutes a by-product of the production of zirconium alloys, and its use in control clusters of reactors allows its value to be exploited.
Hafnium has the advantage over other absorbent materials of having excellent compatibility with the primary coolant and it can therefore be used without cladding. It can be used in the pure state or in a weakly alloyed state with elements such as tin and/or oxygen in order to increase its mechanical characteristics, iron, chromium and/or niobium in order to increase its corrosion resistance, and molybdenum in order to increase its wear resistance. Finally, owing to the difficulties involved in separating hafnium and zirconium, it may comprise up to 2.5% of residual zirconium, preferably a maximum of 1%.
Both for zirconium alloys and for hafnium and its alloys, the chemical properties (such as resistance to various types of corrosion by water from the reactor coolant system) capable of continuing depending on the conditions of use of the alloy: nodule corrosion, general corrosion, corrosion under stress, corrosion under irradiation, etc.) and mechanical properties (tensile strength, elastic limit, flux resistance, friction and wear resistance, etc.) which are desirable are in particular obtained by coordinating the selection of the alloy elements and the contents thereof and the selection of the thermal and thermomechanical processing operations undergone by the alloy ingot during the production of the element, such as spinning, laminating and/or drawing operations, quenching operations, annealing operations, various surface polishing or finishing operations.
The corrosion of a cladding of zirconium alloy becomes evident, for example, owing to the formation of a layer of zircon ZrO2. It brings about a loss in mass of the metal and therefore a reduction in the outside diameter of the metal cladding, and also a substantial modification of the thermal characteristics thereof because zircon is ten times less conductive than the base metal. That oxidation also becomes evident as the formation of hydrogen, a portion of which becomes diffused in the cladding and forms hydrides which embrittle it. It also modifies the friction properties of the surface of the cladding. It further modifies the properties of resistance to wear of the surface of the cladding. It further modifies the nucleation sites of the deposits which crystallise from the heat exchange fluid (which deposits are conventionally referred to as “CRUD”, which is an acronym standing for Chalk River Unidentified Deposit) and may bring about an increase in the quantity of those deposits. It also diminishes the friction conditions of the fluid over the surface, which has a detrimental effect on the pressure drops of the assembly. Finally, it modifies the nucleation of bubbles of vapour, and therefore the thermo-hydraulic behaviour during exchanges between the rod and the heat exchange fluid.
During the production and assembly operations of the fuel assemblies, superficial defects, such as scratches or local microdeformations, are formed on the claddings, produced in particular by friction to which the fuel rods are subjected when they are inserted in the skeleton of the fuel assembly. Those initial superficial defects may aggravate subsequent deterioration of the corrosion or wear behaviour.
Damage may also become evident during operation in the event of contact of the rod and the support elements thereof in the grids, and owing to migrating bodies which become trapped in the grids and strike the cladding. During operation of the nuclear reactor, the coolant water which flows with a high upward speed in the core brings about small oscillating movements of the nuclear fuel rods inside the spacer grids and any foreign bodies jammed in the structure of the assembly. This phenomenon, which is known as “fretting”, will result in friction between the nuclear fuel rods and the contact portions which may bring about wear of the claddings of the nuclear fuel rods. That wear could bring about puncturing of the claddings of the fuel rods, and therefore a release of radioactive materials and gas into the water of the reactor coolant system, which may lead to a shutdown of the reactor for premature removal of the fuel assemblies comprising defective rods. Furthermore, when the cladding is punctured, hydrogen generated by radiolysis of the water of the reactor coolant system which is introduced inside the rod brings about massive hydriding of the cladding which may lead to fractures thereof owing to embrittlement.
A similar phenomenon of fretting exists with regard to the inserts when the BWR control cross is constituted by structural plates which are generally composed of steel, gripping inserts of absorbent material. The wear of the surface layer of oxide of the insert of hafnium allows hydriding thereof and may lead to deformation of the control cross owing to the expansion of the hafnium under the effect of the hydriding action. This may lead to an increase in the volume of the material by 15%. Such deformation is unacceptable because it may bring about an increase to the insertion duration of the control cross, or even jamming thereof.
Furthermore, the guide tubes of the fuel assembly, which are also constructed from zirconium alloy, are exposed to wear owing to friction or fretting on the rods of the control clusters, and vice versa. Such wear of the guide tubes reduces the thickness of metal, where applicable up to puncturing thereof, which involves a risk of causing disruption to the flow of the heat exchange fluid and occurrences of dangerous mechanical weakening during operation or during handling between cycles. Wear of the rod of a cluster or of the control cross is equally harmful, it being possible for mechanical weakening to lead to fracturing of the component involved.
The cladding of the fuel rod is the first confinement barrier of the fission products, the other barriers being constituted by the vessel of the reactor and the concrete vessel.
During normal operation (situations referred to as being of class 1) and malfunctioning operation (situations referred to as being of class 2), the sealing of the cladding with regard to the fission products must therefore be ensured.
During a transient power occurrence corresponding to a situation of class 2, the power locally reached in the fuel may be from two to three times greater than nominal power. That rapid increase in power brings about great expansion of the pellets. The thermal expansion of the pellets brings it to that of the cladding and results in the cladding being placed under traction by the pellets and an increase in the stresses on the internal surface of the cladding, up to a level which may exceed the elastic limit of the material constituting the cladding, thereby bringing about damage to the cladding. Furthermore, that mechanical stress takes place in the presence of an aggressive chemical environment owing to the fission products, such as iodine, released by the fuel during the power occurrence. This is referred to as Pellet Cladding Interaction (PCI), a phenomenon which may lead to rupture of the cladding.
Such rupture of the cladding is not permitted for reasons of safety, because it could lead to the release of fission products in the reactor coolant system of the reactor.
It is conceivable to attempt to solve those different problems by carrying out surface processing operations of the component involved. However, one risk is that, during these processing operations, it may be necessary to bring the component to a relatively high temperature, which would impair the internal metallurgical structure and the mechanical properties of the component obtained following the production operation.
Furthermore, it is still not desirable to process the whole of the surface of the components. Some zones must, for example, be left to allow subsequent welding. It is not always easy, when the surface processing is carried out, to protect the zones to be left. The masking of the zones to be left must be tight, the protective material is at risk of polluting the material to be protected and, before the processed component is put into operation, it is necessary to ensure that all the protective material has been removed. Therefore, the application of conventional surface processing methods often involves implementation difficulties for these reasons.
This is true, for example, of internal coverings of the claddings of fuel rods. It is known to use, in the core of the nuclear reactors, fuel assemblies which comprise rods containing a substance which is highly absorbent with respect to neutrons, such as gadolinium or erbium or zirconium diboride ZrB2. The last one is in particular used in the form of an internal covering of the cladding tubes, which is generally carried out by surface deposition, via methods which are often complex to carry out and which do not ensure constant quality of the covering, with regard to its surface quality, its adhesion and obtaining a defined and constant thickness of the covering. Furthermore, it is often very difficult to avoid the presence of covering in end portions of the tube at which the closure plugs of the fuel rod subsequently have to be welded.
The fuel assembly also comprises planar elements, such as the grid plates or bands. They must have precise mechanical properties, which are obtained by suitable selection of the composition thereof and the processing operations to which they are subjected during production thereof. Again, a surface processing operation may have a positive or negative influence on those mechanical properties, and the development of surface processing operations which are simple and economical to implement and confer improved properties on the component would be highly advantageous.
The sensitivity to wear of the components of hafnium during friction on the guiding elements (guide tubes or housings) requires surface processing. That surface processing can be carried out by diffusion of oxygen at 860° C. for several hours under an argon/oxygen atmosphere. In this case, it is not known how to protect the zones which have to be subsequently processed by machining or formed. Again, there is a need for a surface processing method which, on the one hand, would be used at low temperature and would not therefore excessively change the mechanical properties of the alloy and, on the other hand, would allow efficient masking of the zones which must not be affected by the diffusion of oxygen.
When an improvement to corrosion resistance and/or other mechanical or chemical properties of the component is sought, it is possible to consider using metal or ceramic protective coverings. However, that technology has limits linked to the non-adhesion of the covering and the cracks which may appear in the event of thermal cycles or mechanical stresses.
It is also known that the deposit of CRUD is linked to the precipitation by chemical reaction of colloids present in the reactor coolant system at the surface of the rods, in particular if that surface is rough. Document EP-A-1 486 988 proposes bringing about an alternating arrangement of polished and non-polished zones on the surface of the rods in order to limit adhesion and the growth of the oxide crystals.
It is also possible to carry out processing operations involving diffusion by the thermochemical route of elements such as carbon, nitrogen, boron, oxygen, etc. However, even though they do improve the mechanical strength, some elements may be incompatible with an application to zirconium alloys. The diffusion of carbon and nitrogen impairs the corrosion stability in water at high temperature, and boron is a neutron absorbent which is converted into gaseous helium under irradiation. Furthermore, those processing operations must be carried out at a temperature which is too high to maintain the mechanical properties of the substrate. In this manner, a diffusion of oxygen over a depth of approximately twenty μm could be carried out on an alloy of zirconium prepared according to conventional methods only at a temperature in the order of 800° C. for a few hours, which would inevitably bring about a radical modification of the metallurgical structure of the whole of the material. That would necessitate complete reconsideration of the prior steps of the production method in order to obtain the final properties sought, and it is not at all certain that it would be possible to do so.
In order to carry out a surface diffusion of oxygen over approximately forty μm on hafnium alloys, the temperature or the duration must be even greater (930° C.-3 hours, or 860° C.-6 hours or 800° C.-36 hours), which is incompatible with maintaining high mechanical properties linked with a work-hardened state (recrystallisation temperature in the order of from 700 to 750° C.). The processing operation for diffusion of oxygen on hafnium at 650° C. for 6 hours results only in a layer of oxide having a thickness of the order of 1.5 μm and a subjacent layer of diffusion of oxygen of small thickness (from 3 to 5 μm), which is not enough for wear resistance.